Zirconium alloys with improved corrosion/creep resistance due to final heat treatments

ABSTRACT

Articles, such as tubing or strips, which have excellent corrosion resistance to water or steam at elevated temperatures, are produced from alloys having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weight percent iron, and optionally additional alloy elements selected from the group consisting of tin, chromium, copper, vanadium, and nickel with the balance at least 97 weight percent zirconium, including impurities, where a necessary final heat treatment includes one of i) a SRA or PRXA (15-20% RXA) final heat treatment, or ii) a PRXA (80-95% RXA) or RXA final heat treatment.

CROSS-REFERENCE TO RELATED APPLICATIONS

The present application is a continuation-in-part application whichclaims priority from all the following applications: U.S. patentapplication Ser. No. 12/697,322, filed Feb. 1, 2010, which is adivisional application of U.S. Ser. No. 11/087,844, filed Mar. 23, 2005,which claims priority from U.S. Provisional Application Ser. No.60/555,600, filed Mar. 23, 2004, and Provisional Application Nos.60/564,416, 60/564,417 and 60/564,469, each filed Apr. 22, 2004, thedisclosures of all of which are incorporated herein by reference.

BACKGROUND OF THE INVENTION

1. Field of the Invention

The present invention generally relates to a zirconium based alloyusable for the formation of strips and tubing for use in nuclear fuelreactor assemblies. Specifically, the invention relates to newtechnology that improves the in-reactor corrosion and/or the in-reactorcreep of Zr—Nb based alloys by an essential and critical final heattreatment. The invention was applied to Zr—Nb based alloys that weredeveloped by alloying element additions and exhibit improved corrosionresistance in water based reactors under elevated temperatures.

2. Description of the Prior Art

In the development of nuclear reactors, such as pressurized waterreactors and boiling water reactors, fuel designs impose significantlyincreased demands on all of the fuel components, such as cladding,grids, guide tubes, and the like. Such components are conventionallyfabricated from zirconium-based alloys commercially titled ZIRLO,corrosion resistant alloys that contain about 0.5-2.0 wt. % Nb; 0.9-1.5wt. % Sn; and 0.09-0.11 wt. % of a third alloying element selected fromMo, V, Fe, Cr, Cu, Ni, or W, with the rest Zr, as taught in U.S. Pat.No. 4,649,023 (Sabol et al.). That patent also taught compositionscontaining up to about 0.25 wt. % of the third alloying element, butpreferably about 0.1 wt %. Sabol et al., in “Development of a CladdingAlloy for High Burnup” Zirconium in the Nuclear Industry: EighthInternational Symposium, L. F. Van Swan and C. M. Eucken, Eds., AmericanSociety for Testing and Materials, ASTM STP 1023, Philadelphia, 1989.pp. 227-244, reported improved properties of corrosion resistance forZIRLO (0.99 wt. % Nb, 0.96 wt. % Sn, 0.10 wt. % Fe, remainder primarilyzirconium) relative to Zircaloy-4.

There have been increased demands on such nuclear core components, inthe form of longer required residence times and higher coolanttemperatures, both of which cause increase alloy corrosion. Theseincreased demands have prompted the development of alloys that haveimproved corrosion and hydriding resistance, as well as adequatefabricability and mechanical properties. Further publications in thisarea include U.S. Pat. No. 5,940,464; 6,514,360 (Mardon et al. and Jeonget al.) and Reexamination Certificate U.S. Pat. No. 5,940,464 C1 (bothMardon et al.), and the paper “Advanced Cladding Material for PWRApplication: AXIOM™”, Pan et al., Proceedings of 2010 LWR FuelPerformance/Top Fuel/WRFPM, Orlando, Fla. 09/26-29/2010 (“technicalpaper”).

Mardon et al. taught zirconium alloy tubes for forming the whole orouter portion of a nuclear fuel cladding or assembly guide tube having alow tin composition: 0.8-1.8 wt. % Nb; 0.2-0.6 wt. % Sn, 0.02-0.4 wt. %Fe, with a carbon content of 30-180 ppm, a silicon content of 10-120 ppmand an oxygen content of 600-1800 ppm, with the rest Zr. Jeong et al.taught a niobium containing zirconium alloy for high burn-up nuclearfuel application containing Nb, Sn, Fe, Cr, Zr with possible addition ofCu. The Pan et al. “technical paper” lists Alloys listed as X1, X4, X5,X5A, but deliberately only very generally describes the actualcomposition weight percentages, being very vague in this regard. Pan etal. reports tensile strength, elongation and creep test data, and showsmicrographs and in-reactor corrosion and oxide thickness data.

Aqueous corrosion in zirconium alloys is a complex, multi-step process.Corrosion of the alloys in reactors is further complicated by thepresence of an intense radiation field which may affect each step in thecorrosion process. In the early stages of oxidation, a thin compactblack oxide film develops that is protective and retards furtheroxidation. This dense layer of zirconia exhibits a tetragonal crystalstructure which is normally stable at high pressure and temperature. Asthe oxidation proceeds, the compressive stresses in the oxide layercannot be counterbalanced by the tensile stresses in the metallicsubstrate and the oxide undergoes a transition. Once this transition hasoccurred, only a portion of the oxide layer remains protective. Thedense oxide layer is then renewed below the transformed oxide. A newdense oxide layer grows underneath the porous oxide. Corrosion inzirconium alloys is characterized by this repetitive process of growthand transition. Eventually, the process results in a relatively thickouter layer of non-protective, porous oxide. There have been a widevariety of studies on corrosion processes in zirconium alloys. Thesestudies range from field measurements of oxide thickness on irradiatedfuel rod cladding to detailed micro-characterization of oxides formed onzirconium alloys under well-controlled laboratory conditions. However,the in-reactor corrosion of zirconium alloys is an extremelycomplicated, multi-parameter process. No single theory has yet been ableto completely define it.

Corrosion is accelerated in the presence of lithium hydroxide. Aspressurized water reactor (PWR) coolant contains lithium, accelerationof corrosion due to concentration of lithium in the oxide layer must beavoided. Several disclosures in U.S. Pat. Nos. 5,112,573 and 5,230,758(both Foster et al.) taught an improved ZIRLO composition that was moreeconomically produced and provided a more easily controlled compositionwhile maintaining corrosion resistance similar to previous ZIRLOcompositions. It contained 0.5-2.0 wt. % Nb; 0.7-1.5 wt. % Sn; 0.07-0.14wt. % Fe and 0.03-0.14 wt. % of at least one of Ni and Cr, with the restZr. This alloy had a 520° C. high temperature steam weight gain at 15days of no more than 633 mg/dm². U.S. Pat. No. 4,938,920 to Garzarolliteaches a composition having 0-1 wt. % Nb; 0-0.8 wt. % Sn, and at leasttwo metals selected from iron, chromium and vanadium. However,Garzarolli does not disclose an alloy that had both niobium and tin,only one or the other.

Sabol et al. in “In-Reactor Corrosion Performance of ZIRLO andZircaloy-4,” Zirconium in the Nuclear Industry: Tenth InternationalSymposium, A. M. Garde and E. R. Bradley Eds., American Society forTesting and Materials, ASTM STP 1245, Philadelphia 1994, pp. 724-744,demonstrated that, in addition to improved corrosion performance, ZIRLOmaterial also has greater dimensional stability (specifically,irradiation creep and irradiation growth) than Zircaloy-4. Morerecently, U.S. Pat. No. 5,560,790 (Nikulina et al.) taughtzirconium-based materials having high tin contents where themicrostructure contained Zr—Fe—Nb particles. The alloy compositioncontained: 0.5-1.5 wt. % Nb; 0.9-1.5 wt. % Sn; 0.3-0.6 wt. % Fe, withminor amounts of Cr, C, 0 and Si, with the rest Zr.

While these modified zirconium based compositions are claimed to provideimproved corrosion resistance as well as improved fabricationproperties, economics have driven the operation of nuclear power plantsto higher coolant temperatures, higher burn-ups, higher concentrationsof lithium in the coolant, longer cycles, and longer in-core residencetimes that have resulted in increased corrosion duty for the cladding.Continuation of this trend as burn-ups approach and exceed 70,000MWd/MTU will require further improvement in the corrosion properties ofzirconium based alloys. The alloys of this invention provide suchcorrosion resistance.

Another potential way to increase corrosion resistance is through themethod of forming of the alloy itself. To form alloy elements into atubing or strip, ingots are conventionally vacuum melted and betaquenched, and thereafter formed into an alloy through a gauntlet ofreductions, intermediate anneals, and final anneals, wherein theintermediate anneal temperature is typically above 1105° F. for at leastone of the intermediate anneals. In U.S. Pat. No. 4,649,023 to Sabol etal., the ingots are extruded into a tube after the beta quench, betaannealed, and thereafter alternatively cold worked in a pilger mill andintermediately annealed at least three times. While a broad range ofintermediate anneal temperatures are disclosed, the first intermediateanneal temperature is preferably 1112° F., followed by a laterintermediate anneal temperature of 1076° F. The beta annealing steppreferably uses temperatures of about 1750° F. Foster et al., in U.S.Pat. No. 5,230,758, determined the formability and steam corrosion forintermediate anneal temperatures of 1100° F., 1250° F., and 1350° F. Anincrease in intermediate anneal temperature is associated with anincrease in both formability and corrosion resistance. U.S. Pat. No.5,887,045 to Mardon et al. discloses an alloy forming method having atleast two intermediate annealing steps carried out between 1184° to1400° F.

Note that the prior art for corrosion improvement summarized aboveinvolves alloying element additions and different intermediate annealtemperatures, but, notably, not the final anneal heat treatmenttemperature. Rudling et al., in, “Corrosion Performance of Zircaloy-2and Zircaloy-4 PWR Fuel Cladding,” Zirconium in the Nuclear Industry:Eight International Symposium, ASTM STP 1023, L. F. Van Swam and C. M.Eucken, eds. American Society for Testing and Materials, Philadelphia,1989, pp. 213-226, reported that Zr-4 fuel rod cladding fabricated fromthe same ingot with final heat treatments of stress-relieved (SRA) andfully recrystallized (RXA) exhibited similar oxide thickness corrosion(see Table 1).

TABLE 1 Post irradiation oxide thickness of Zr-4 cladding after 1-cycleof irradiation. Final Heat 4 Rod Average of the Maximum Treatment OxideThickness (μm) SRA 12 +/− 1 RXA 10 +/− 1

Foster et al., in U.S. Pat. No. 5,125,985, presents a straightforwardmethod of controlling the creep by use of the final area reduction andintermediate anneal temperature. A decrease in final area reductiondecreases creep, and an increase in intermediate anneal temperaturedecreases creep. In different applications, the in-reactor creep can bemore important than in-reactor corrosion. One such example is fuel rodscontaining fuel pellets coated with ZrB₂. ZrB₂ is a neutron absorber.When neutrons are absorbed, He gas is released which increases the rodinternal pressure. In this case, creep resistant cladding is necessaryso that the fuel/cladding gap remains closed. A closed fuel/cladding gapensures that the fuel temperatures do not increase due to the formationof a He gas gap between the fuel and cladding. The new technologypresented below in the Summary of the Invention will show that eitherthe cladding corrosion or the cladding in-reactor creep may be improvedby the final heat treatment.

A further issue in nuclear reactors is corrosion of welds utilized in anuclear fuel assembly. In a typical fuel rod, nuclear fuel pellets areplaced within the cladding, which is enclosed by end caps on either endof the cladding, such that the end caps are welded to the cladding. Theweld connecting the end caps to the cladding, however, generallyexhibits corrosion to an even greater extent than the cladding itself,usually by a factor of two over non-welded metal. Rapid corrosion of theweld creates an even greater safety risk than the corrosion ofnon-welded material, and its protection has previously been ignored. Inaddition, grids have many welds and the structural integrity depends onadequate weld corrosion resistance.

Thus, there continually remains a vital need, even in this late stage ofnuclear power development, for novel zirconium cladding alloys thatexhibit improved corrosion resistance and improved in-reactorirradiation creep resistance over known alloys in the field, andimproved welds for holding end caps on claddings and for joining gridstraps that likewise exhibit increased corrosion resistance. And, as canbe seen, these cladding art patents and papers provide an extremelycompact art area, where only very minor changes have shown, afterextended testing, major and dramatic improvements. Thus, minorimprovements can easily establish patentability in this specific area.

Accordingly, an object of the present invention is to provide Zr—Nballoys with improved corrosion resistance and/or improved in-reactorirradiation creep resistance through the selection of a specific typecombination of final heat treatment.

New technology presented below in the Summary of the Invention, andelsewhere in the specification following, will show that the in-reactorcorrosion is, in part, unexpectedly dependent on the specific type offinal heat treatment.

SUMMARY OF THE INVENTION

The Zr—Nb alloys of this invention have improved alloy chemistry,improved weld corrosion resistance, and improved method of formation ofalloys having reduced intermediate anneal temperatures during formationof the alloys.

The new technology showing the effect of an essential and critical finalheat treatment (and the final microstructure) on the in-reactorcorrosion of Zr—Nb—Sn—Fe type alloys is presented in FIGS. 1 and 2. FIG.1 shows the in-reactor oxide thickness corrosion data for 0.77 weight %Sn ZIRLO irradiated for 1, 2 and 3 cycles in the Vogtle Unit 2 PWR. Allof the cladding was fabricated from the same ingot and receivedidentical processing except for the final heat treatment. The claddingwas given 3 different final anneal heat treatments of stress reliefannealed (“SRA”), partially recrystallized (“PRXA”) and fullyrecrystallized (“RXA”). The amount of recrystallization in the PRXA heattreatment was about 15-20%.

A generic composition useful in this invention, to provide unexpectedresults in corrosion resistance and/or in-reactor irradiation creepresistance, is an alloy comprising, as in Claim 1:

0.2 to 1.5 weight percent niobium,

0.01 to 0.6 weight percent iron,

and additional alloying elements selected from the group consisting of:

-   -   0.0 to 0.8 weight percent tin    -   0.0 to 0.5 weight percent chromium    -   0.0 to 0.3 weight percent copper    -   0.0 to 0.3 weight percent vanadium    -   0.0 to 0.1 weight percent nickel, with

the balance at least 97 weight percent zirconium, including impurities,wherein said alloy is characterized in that it has improved corrosionresistance properties due to a final heat treatment selected from one ofi) SRA or PRXA (15-20% RXA) providing low corrosion resistance; or ii)RXA or PRXA (80-95% RXA) providing low creep rate. Impurities mean lessthan 60 ppm or 0.006 wt. %.

Other more specific compositions are set forth in the specification andclaims.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1, which, in part, best illustrates the invention, is a graphdepicting the oxide thickness due to corrosion versus the roomtemperature yield stress and associated final alloy microstructure of0.77 Low-Sn ZIRLO;

FIG. 2, which, in part, best illustrates the invention, is a graphdepicting the oxide thickness due to corrosion versus the roomtemperature yield stress and the associated final alloy microstructureof Standard ZIRLO;

FIG. 3 is a graph depicting the in-reactor irradiation creep rate versusthe room temperature as-fabricated yield stress and the associated finalalloy microstructure of 0.77 Sn ZIRLO;

FIG. 4, which, in part, best illustrates the invention, is a graphdepicting oxide thickness due to corrosion as a function of the burn-upfor Standard ZIRLO, Optimized ZIRLO and Alloys X1, X4 and X5;

FIG. 5A is a process flow diagram of a method for forming zirconiumalloy tubing;

FIG. 5B is a process flow diagram of a method for forming zirconiumalloy strips;

FIG. 6 is a graph depicting the 680° F. water test weight gain ofStandard ZIRLO as a function of autoclave exposure time for materialprocessed with intermediate anneal temperatures of 1085° and 1030° F.;

FIG. 7 is a graph depicting the 680° F. water test weight gain of AlloyX1 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 8 is a graph depicting the 680° F. water test weight gain of AlloyX4 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 9 is a graph depicting the 680° F. water test weight gain of AlloyX5 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 10 is a graph depicting the 680° F. water test weight gain of AlloyX6 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 11 is a graph depicting the 800° F. steam test weight gain ofStandard ZIRLO as a function of autoclave exposure time for materialprocessed with intermediate anneal temperatures of 1085° and 1030° F.;

FIG. 12 is a graph depicting the 800° F. steam test weight gain of AlloyX1 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 13 is a graph depicting the 800° F. steam test weight gain of AlloyX4 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 14 is a graph depicting the 800° F. steam test weight gain of AlloyX5 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.;

FIG. 15 is a graph depicting the 800° F. steam test weight gain of AlloyX6 as a function of autoclave exposure time for material processed withintermediate anneal temperatures of 1085° and 1030° F.; and

FIG. 16 is a graph comparing the 800° F. steam weight gain for StandardZIRLO strip processed with low temperature intermediate and final annealtemperatures.

DESCRIPTION OF THE PREFERRED EMBODIMENTS

Referring now to the drawings; FIG. 1 very importantly shows that theoxide thickness depends on the final heat treatment. FIG. 1 presents thecorrosion of 0.77 Sn ZIRLO. All of the cladding was fabricated from thesame ingot and received identical processing except for the final heattreatment. The cladding was given three final heat treatments of SRA,PRXA and RXA. The highest corrosion (highest oxide thickness) wasexhibited by cladding with the RXA —fully recrystallized—final heattreatment. Significantly lower corrosion was exhibited by cladding withboth SRA and PRXA (15% to 20%) final heat treatments.

FIG. 2 very importantly shows the in-reactor oxide thickness corrosiondata for Standard ZIRLO (1.02 weight % Sn) irradiated for 1, 2 and 3cycles in the Vogtle Unit 2 PWR. All of the cladding was fabricated fromthe same ingot and received identical processing except for the finalheat treatment. The cladding was given 2 different final anneal heattreatments of SRA and RXA. FIG. 2, very importantly, shows that theoxide thickness depends on the final heat treatment as exhibited by the0.77 weight % Sn ZIRLO data in FIG. 1. The highest corrosion (highestoxide thickness) was exhibited by cladding with the RXA final heattreatment. Significantly lower corrosion was exhibited by cladding withthe SRA final heat treatment.

As discussed above, depending on the application, improved in-reactorcreep resistance can be as important as improved corrosion resistance.The in-reactor creep is also dependent on the final heat treatment. FIG.3, very importantly, presents the in-reactor steady state creep rate for0.77 weight % Sn ZIRLO (which is Claim 1 of this invention) irradiatedfor 1, 2 and 3 cycles in the Vogtle Unit 2 PWR (see paragraph 13). FIG.3 shows that the highest in-reactor creep resistance (that is, thelowest in-reactor creep rate) is exhibited by cladding with a RXA finalheat treatment. The lowest in-reactor creep resistance (that is, thehighest in-reactor creep rate) is exhibited by cladding with a SRA finalheat treatment. Intermediate in-reactor creep resistance is exhibited bythe PRXA final heat treatment. Thus, both SRA and PRXA are effective inthis regard with RXA the best.

Hence, the effect of final heat treatment on in-reactor creep isopposite that of in-reactor corrosion. As a result, the cladding may beoptimized for either maximum improved in-reactor corrosion resistancewith a SRA or PRXA (15-20% RXA) final heat treatment, or maximumimproved in-reactor creep resistance with a final PRXA (80-95% RXA) orRXA heat treatment.

In more substantial detail, each of these “terms,” RXA, PRXA, SRA, etc.is defined as:

-   -   SRA means—heat treatment where the microstructure is        stress-relief annealed.    -   RXA means—heat treatment where the microstructure is fully        recrystallized.    -   PRXA (15-20% RXA) means—heat treatment where 15-20% of the        microstructure is recrystallized and 80-85% of the        microstructure is stress relief annealed.    -   PRXA (80-95% RXA) means—heat treatment where 80-95% of the        microstructure is recrystallized and 5-20% of the microstructure        is stress relief annealed.

Note that the above SRA, PRXA and RXA designations represent moredetailed descriptions of the final heat treatment process methods. Itshould be clear that this art area is not an area in patent filing wherebroad conclusions are suggestive of improved alloys within broad ranges;where, for example, 0.4 to 1.5 weight percent niobium and 0.1 to 0.8weight percent tin, should be considered taught or obvious in view of ateaching of 0.0 to 3.0 weight percent niobium and 0.1 to 3.5 weightpercent tin. As shown in FIG. 4, standard Zirlo compared to compositionsX4 and X5 shows the dramatic difference a few tenths of weight percentelements make in this area:

Standard Zirlo: 0.5-2 wt % Nb; 0.9-1.5 wt. % Sn

X4: 1 wt. % Nb; 0 wt. % Sn, etc. or

X5: 0.7 wt. % Nb; 0.3 wt. % Sn, etc.;

where these seemingly reduced and very important minor changes incomponent elements

provide extraordinarily improved oxide thickness. Specifically, at aburnup of 70

GWd/MTU, the oxide thickness is reduced at least by a factor of 3.5.

FIG. 4, very dramatically, illustrates at 75 GWd/MTU a range of oxidethickness of about 35-40 micrometer for alloy X1, and a range of about16 to 26 micrometers for alloys X4 and X5, all showing criticalimprovements relative to standard ZIRLO.

A further object of the present invention is to provide a zirconiumbased alloy for use in an elevated temperature environment of a nuclearreactor, the alloy having 0.2 to 1.5 weight percent niobium, 0.01 to 0.6weight percent iron, and additional alloy elements selected from 0.0 to0.8 weight percent tin, 0.0 to 0.5 weight percent chromium, 0.0 to 0.3weight percent copper, 0.0 to 0.3 weight percent vanadium, 0.0 to 0.1weight percent nickel, the remainder at least 97 weight percentzirconium, including impurities. Further descriptions of vastly improvedalloys X1, X4 and X5 follow.

Alloy X4: A further object of the present invention is to provide azirconium based alloy (denoted as Alloy X4) for use in an elevatedtemperature environment of a nuclear reactor, the alloy having 0.6 to1.5 weight percent niobium, 0.02 to 0.3 weight percent Cu, 0.01 to 0.1weight percent iron, 0.15 to 0.35 weight percent chromium, the balanceat least 97 weight percent zirconium, including impurities (Claims9-12).

Alloy X5: A further object of the present invention is to provide azirconium based alloy (denoted as Alloy X5), the alloy having 0.2 to 1.5weight percent niobium, 0.25 to 0.45 weight percent iron, 0.05 to 0.4weight percent tin, 0.15 to 0.35 weight percent chromium, 0.01 to 0.1weight percent nickel, the balance at least 97 weight percent zirconium,including impurities (Claims 13-16).

Alloy X1: A further object of the invention is to provide a zirconiumbased alloy (denoted as Alloy X1), the alloy having 0.4 to 1.5 weightpercent niobium, 0.05 to 0.4 weight percent tin, 0.01 to 0.1 weightpercent iron, 0.02 to 0.3 weight percent copper, 0.12 to 0.3 weightpercent vanadium, 0.0 to 0.5 weight percent chromium, the balance atleast 97 weight percent zirconium, including impurities (Claims 4-8).

Alloy X6: A further specific object of the invention is to provide azirconium based alloy (denoted as Alloy X6 and referred to as“Optimized” ZIRLO), shown in FIG. 4, the alloy having 0.4 to 1.5 weightpercent niobium, 0.1 to 0.8 weight percent tin, 0.01 to 0.6 weightpercent iron, 0.0 to 0.5 weight percent chromium, the balance at least97 weight percent zirconium, including impurities (Claims 17-22). Thisalloy is still vastly superior to standard ZIRLO.

The final heat treatment of Alloy X1 is PRXA (−80% RXA), which isassociated with maximum, improved (low) in-reactor creep resistance. Inaddition, note that the corrosion resistance of Alloy X1 issignificantly increased relative to Standard ZIRLO, by a factor of 2.2at a burn-up of 70 GWd/MTU (see FIG. 4), because of decreased Sn and theaddition of Cu. Further, if the amount of RXA in the PRXA final heattreatment of Alloy X1 is decreased to about 15-20%, the corrosionresistance of Alloy X1 would be further improved.

The final heat treatment of Alloy X4 is PRXA (−80% RXA) which isassociated with maximum improved in-reactor creep resistance. At aburn-up of 70 GWd/MTU, the corrosion resistance of Alloy X4 is increasedbe a factor of about 3.5 (see FIG. 4) relative to Standard ZIRLO. Notethat the corrosion resistance of Alloy X4 is significantly increasedrelative to Standard ZIRLO because of decreased Sn and the additions ofCu and Cr. In addition, if the amount of RXA in the PRXA final heattreatment of Alloy X4 is decreased to about 15-20% PRXA (15-20% RXA),the corrosion resistance of Alloy X4 would by further improved.

The final heat treatment of Alloy X5 is PRXA (−50% RXA), which isconsidered to be intermediate between maximum improved in-reactor creepresistance and maximum improved in-reactor corrosion resistance. FIG. 4shows that at a burn-up of 70 GWd/MTU, the corrosion resistance of AlloyX5 is increased be a factor of about 3.0 relative to Standard ZIRLO.Note that the corrosion resistance of Alloy X5 is significantlyincreased relative to Standard ZIRLO because of decreased Sn, increasedFe and the addition of Cr.

A sequence of steps for forming a cladding, strip, tube or like objectknown in the art from an alloy of the present invention is shown inFIGS. 5A and 5B. To create tubing for cladding, as shown in FIG. 5A,compositional zirconium based alloys were fabricated from vacuum meltedingots or other like material known in the art. The ingots werepreferably vacuum arc-melted from sponge zirconium with a specifiedamount of alloying elements. The ingots were then forged into a materialand thereafter β-quenched. β-quenching is typically done by heating thematerial (also known as a billet) up to its β-temperature, betweenaround 1273 to 1343K. The quenching generally consists of quicklycooling the material by water. The β-quench is followed by extrusion.Thereafter, the processing includes cold working the tube-shell by aplurality of cold reduction steps, alternating with a series ofintermediate anneals at a set temperature. The cold reduction steps arepreferably done on a pilger mill. The intermediate anneals are conductedat a temperature in the range of 960-1125° F. The material may beoptionally re-β-quenched prior to the final and foamed into an articlethere-from. The final heat treatment discussed previously is also shown.

For tubing, a more preferred sequence of events after extrusion includesinitially cold reducing the material in a pilger mill, an intermediateanneal with a temperature of about 1030 to 1125° F., a second coldreducing step, a second intermediate anneal within a temperature rangeof about 1030° to 1070° F., a third cold reducing step, and a thirdintermediate anneal within a temperature range of about 1030° to 1070°F. The reducing step prior to the first intermediate anneal is a tubereduced extrusion (TREX), preferably reducing the tubing about 55%.Subsequent reductions preferably reduce the tube about 70-80%.

Each reduction pass on the pilger mill is preferred to reduce thematerial being formed by at least 51%. The material then preferably goesthrough a final cold reduction. The material is then processed with afinal anneal at temperatures from about 800-1300° F.

To create strip, compositional zirconium based alloys were fabricatedfrom vacuum melted ingots or other like material known in the art. Theingots were preferably arc-melted from sponge zirconium with a specifiedamount of alloying elements. The ingots were then forged into a materialof rectangular cross-section and thereafter β-quenched. Thereafter, theprocessing as shown in FIG. 5B, includes a hot rolling step after thebeta quench, cold working by one or a plurality of cold rolling andintermediate anneal steps, wherein the intermediate anneal temperatureis conducted at a temperature from about 960-1105° F. The material thenpreferably goes through a final pass and anneal, wherein the finalanneal temperature is in the range of about 800-1300° F. The final heattreatment discussed previously is also shown.

A more preferred sequence to create the alloy strip includes anintermediate anneal temperature within a range of about 1030 to 1070° F.Further, the pass on the mill preferably reduces the material beingformed by at least 40%.

The corrosion resistance was found to improve with intermediate annealsalso that were consistently in the range of 960-1105° F., mostpreferably around 1030-1070° F., as opposed to typical prior annealtemperatures that are above the 1105° F. for at least one of thetemperature anneals. As shown in FIGS. 6-10, a series of preferred alloyembodiments of the present invention were tested for corrosion in a 680°F. water autoclave and measured for weight gain. Tubing material wasfabricated from the preferred embodiments of alloys of the presentinvention, referenced as Alloys X1, X4, X5 and X6, and placed in the680° F. water autoclave. Data were available for a period of 100 days.Corrosion resistance measured in 680° F. water autoclaves for long termexposure have previously been found to correlate to corrosion resistancedata of like alloys placed in-reactor. The preferred composition ofthese embodiments, further discussed below, are shown in Table 2. Thepreferred ranges of the compositions are presented in Table 3.

TABLE 2 Alloy Preferred Composition, by weight percentage X1Zr—0.7Nb—0.3Sn—0.12Cu—0.18V—0.05Fe X1 Zr—1.0Nb—0.3Sn—0.12Cu—0.18V—0.05FeX1 + Cr Zr—0.7Nb—0.3Sn—0.12Cu—0.18V—0.05Fe—0.2Cr X1 + CrZr—1.0Nb—0.3Sn—0.12Cu—0.18V—0.05Fe—0.2Cr X4Zr—1.0Nb—0.05Fe—0.25Cr—0.08Cu X5 Zr—0.7Nb—0.3Sn—0.3Fe—0.25Cr—0.05Ni X6Zr—1.0Nb—0.65Sn—0.1Fe X6 + Cr Zr—1.0Nb—0.65Sn—0.1Fe—0.2Cr

TABLE 3 Alloy Preferred Composition Ranges, by weight percentage X1 Zr;0.4-1.5Nb; 0.05-0.4Sn; 0.01-0.1Fe; 0.02-0.3Cu; 0.12-0.3V X1 − Cr Zr;0.4-1.5Nb; 0.05-0.4Sn; 0.01-0.1Fe; 0.02-0.3Cu; 0.12-0.3V; 0.05-0.5Cr X4Zr; 0.6-1.5Nb; 0.01-0.1Fe; 0.02-0.3Cu; 0.15-0.35Cr X5 Zr; 0.2-1.5Nb;0.05-0.4Sn; 0.25-0.45Fe; 0.15-0.35Cr; 0.01-0.1Ni X6 Zr; 0.4-1.5Nb;0.14-0.8Sn; 0.01-0.6Fe X6 + Cr Zr; 0.4-1.5Nb; 0.1-0.8Sn; 0.01-0.6Fe;0.05-0.5Cr

In order to evaluate the effect of intermediate anneal temperature oncorrosion/oxidation, tubing of Standard ZIRLO and Alloys X1, X4 and X5were processed with intermediate anneal temperatures of 1030° and 1085°F. The alloys of the invention were tested for corrosion resistance bymeasuring the weight gain over a period of time, wherein the weight gainis mainly attributable to an increase of oxygen (the hydrogen pickupcontribution to the weight gain is relatively small and may beneglected) that occurs during the corrosion process. In general,corrosion related weight gain starts quickly and then the rate decreaseswith increasing time. This initial corrosion/oxidation process is termedas pre-transition corrosion. After a period of time, the corrosion rateincreases, approximately linearly with time. This corrosion/oxidationphase is termed post-transition or rapid corrosion. As would beexpected, alloys with greater corrosion resistance have lower corrosionrates in the pre- and post-transition phases.

FIGS. 6-10 present 680° F. water corrosion test data. As can be seen inFIGS. 6-10, the weight gain associated with tubing processed with 1030°F. intermediate anneal temperatures was less than for higherintermediate anneal temperatures. Further, the weight gains for AlloysX1, X4, X5 and X6 in FIGS. 7-10 were less than that of Standard ZIRLO inFIG. 6. Thus, as the modified alloy compositions and the lowerintermediate anneal temperatures exhibit reduced weight gain, andreduced weight gain is correlated with increased corrosion resistance,increased corrosion resistance is directly correlated with the modifiedalloy compositions and the lower intermediate anneal temperature of theinvention. The chemistry formulation of the alloys is correlated withincreased corrosion resistance. All of the weight gains from the 680° F.water autoclave testing presented in FIGS. 6-10 are in thepre-transition phase. Although the improvement in the 680° F. waterautoclave corrosion weight gain due to lowering of the intermediateanneal temperature appears to be small in view of FIGS. 6-10, theimprovement of in-reactor corrosion resistance is expected to be higherthan shown by the 680° F. water autoclave data because of in-reactorprecipitation of second phase particles in these Zr—Nb alloys and athermal feedback from a lower oxide conductivity due to lower oxidethickness. Such second phase particle precipitation only occursin-reactor and not in autoclave testing.

In order to evaluate the effect of intermediate anneal temperature inpost-transition corrosion, an 800° F. steam autoclave test wasperformed, as shown in FIGS. 11-15. The test was performed forsufficient time to achieve post-transition corrosion. Post transitioncorrosion rates generally began after a weight gain of about 80 mg/dm².Alloys X1, X4, X5 and Standard ZIRLO were processed using intermediateanneal temperatures of 1030° and 1085° F. Alloy X6 (Optimized Zirlo)tubing was processed using intermediate anneal temperatures of 1030° and1105° F. The tubing was placed in an 800° F. steam autoclave for aperiod of about 110 days. FIGS. 11-15 show that the post-transitionweight gains of the alloys processed at the intermediate annealtemperature of 1030° F. are less than for alloy materials processed atthe higher temperatures of 1085° or 1105° F. Further, the weight gainfor Alloys X1, X4, X5 and X6 (Optimized Zirlo) of FIGS. 12-15 are lessthan those of the prior disclosed Standard ZIRLO presented in FIG. 11.Thus, the low intermediate anneal temperatures provide substantialimprovements over the prior art as it provides a significant advantagein safety, by protecting cladding or the grids from corrosion, in cost,as replacement of the fuel assemblies can be done less often, andthrough efficiency, as the less corroded cladding better transmits theenergy of the fuel rod to the coolant.

Standard ZIRLO strip was processed with intermediate anneal temperaturesof 968° and 1112° F. The material was tested for corrosion resistance bymeasuring the weight gain over a period of time, wherein the weight gainis mainly attributable to an increase of oxygen (the hydrogen pickupcontribution to the weight gain is relatively small and may beneglected) that occurs during the corrosion process. The low temperaturestrip was processed with an intermediate anneal temperature of 968° F.and a final anneal temperature of 1112° F. The standard strip wasprocessed with an intermediate anneal temperature of 1112° F. and afinal anneal temperature of 1157° F. FIG. 16 shows that the lowtemperature processed material exhibits significantly lowercorrosion/oxidation than the higher temperature processed material.

The zirconium alloys of the present invention provide improved corrosionresistance through the chemistry of new alloy combinations. The alloysare generally formed into cladding (to enclose fuel pellets) and strip(for spacing fuel rods) for use in a water based nuclear reactor. Thealloys generally include 0.2 to 1.5 weight percent niobium, 0.01 to 0.6weight percent iron, and additional alloying elements selected from thegroup consisting of 0.0 to 0.8 weight percent tin, 0.0 to 0.5 weightpercent chromium, 0.0 to 0.3 weight percent copper, 0.0 to 0.3 weightpercent vanadium and 0.01 to 0.1 weight percent nickel. The balances ofthe alloys are at least 97 weight percent zirconium, includingimpurities. Impurities may include about 900 to 1500 ppm of oxygen.

A first embodiment of the present invention is a zirconium alloy having,by weight percent, about 0.4-1.5% Nb; 0.05-0.4% Sn, 0.01-0.1% Fe,0.02-0.3% Cu, 0.12-0.3% V, 0.0-0.5% Cr and at least 97% Zr includingimpurities, hereinafter designated as Alloy X1. This embodiment, and allsubsequent embodiments, should have no more than 0.50 wt. % additionalother component elements, preferably no more than 0.30 wt. % additionalother component elements, such as nickel, chromium, carbon, silicon,oxygen and the like, and with the remainder Zr. Chromium is an optionaladdition to Alloy X1. Wherein chromium is added to Alloy X1, the alloyis hereinafter designated as Alloy X1+Cr.

Alloy X1 was fabricated into tubing and its corrosion rate was comparedto that of a series of alloys likewise fabricated into tubing, includingZIRLO-type alloys and Zr—Nb compositions. The results are shown in FIG.4. FIG. 4 shows that the in-reactor corrosion resistance of Alloy X1 isincreased by a factor of 2.2 relative to Standard ZIRLO. The chemistryformulations of Alloy X1 provide substantial improvement over the priorart as it relates to corrosion resistance in a nuclear reactor.

A second embodiment of the present invention is a zirconium alloyhaving, by weight percent, about, about 0.6-1.5% Nb; 0.01-0.1% Fe,0.02-0.3% Cu, 0.15-0.35% Cr and at least 97% Zr, hereinafter designatedas Alloy X4. FIG. 4 shows that the in-reactor corrosion resistance ofAlloy X4 is increased by a factor of 3.5 relative to Standard ZIRLO. Apreferred composition of Alloy X4 has weight percent ranges for thealloy with about 1.0% Nb, about 0.05% Fe, about 0.25% Cr, about 0.08%Cu, and at least 97% Zr.

The preferred Alloy X4 was fabricated into tubing and its corrosion ratewas compared with the corrosion rate of Standard ZIRLO. The chemistryformulations of Alloy X4, like Alloy X1, provides substantialimprovements over the prior art as it relates to corrosion resistance ina nuclear reactor.

A third embodiment of the present invention is a zirconium alloy having,by weight percent, about 0.2-1.5% Nb; 0.05-0.4% Sn, 0.25-0.45% Fe,0.15-0.35% Cr, 0.01-0.1% Ni, and at least 97% Zr, hereinafter designatedas Alloy X5. This composition should have no more than 0.5 wt. %additional other component elements, preferably no more than 0.3 wt. %additional other component elements, such as carbon, silicon, oxygen andthe like, and with the remainder Zr.

A preferred composition of Alloy X5 has weight percent values for thealloy with about 0.7% Nb; about 0.3% Sn, about 0.35% Fe, about 0.25% Cr,about 0.05% Ni, and at least 97% Zr.

The preferred embodiment of Alloy X5 was fabricated into tubing and itscorrosion rate was compared to that of a series of alloys likewisefabricated into tubing. FIG. 4 shows that the in-reactor corrosionresistance of Alloy X5 is increased by a factor of 3.0 relative toStandard ZIRLO.

The chemistry formulations of Alloy X5 provide substantial improvementover the prior art as it relates to corrosion resistance in a nuclearreactor.

Another embodiment of the invention is a low-tin ZIRLO alloy designatedas Alloy X6 (“Optimized Zirlo”). FIG. 4 shows that the corrosionin-reactor resistance of Alloy X6 is increased by a factor of 1.5relative to Standard ZIRLO. The reduction of tin increases the corrosionresistance. Tin, however, increases the in-reactor creep strength, andtoo small an amount of tin makes it difficult to maintain the desiredcreep strength of the alloy. Thus, the optimum tin of this alloy mustbalance these two factors. As a result, this embodiment is a low-tinalloy essentially containing, by weight percent, 0.4-1.5% Nb; 0.1-0.8%Sn, 0.01-0.6% Fe, and the balance at least 97% Zr, including impurities,hereinafter designated as Alloy X6. A preferred composition of Alloy X6has weight percent ranges of about 1.0% Nb, about 0.65% Sn, about 0.1%Fe, and at least 97% Zr, including impurities.

Tin may be decreased if other alloy elements are included to replace thestrengthening effect of tin. A second preferred embodiment of Alloy X6(“Optimized Zirlo”) has generally the same weight percentages plus0.05-0.5% Cr, hereinafter designated as Alloy X6+Cr. A preferredembodiment of Alloy X6+Cr has about 1.0% Nb, about 0.65% Sn, about 0.1%Fe and about 0.2% Cr.

Alloy X6 provides substantial improvements in comparison to StandardZIRLO over the prior art as it relates to corrosion resistance in anuclear reactor.

Weld-Corrosion Resistance In a typical nuclear fuel assembly largenumbers of fuel rods are included. In each fuel rod nuclear fuel pelletsare placed within cladding tubes that are sealed by end caps such thatthe end caps are welded to the cladding. The end cap-cladding weld,however, is susceptible to corrosion to an even greater extent than thenon-welded cladding itself, usually by a factor of two.

Zirconium alloys that include chromium show increased weld corrosionresistance. Thus, the addition of chromium in a zirconium alloy includessubstantial advancement over prior zirconium alloys that do not includechromium.

Multiplicities of alloys were tested for their effect on weld corrosion,as shown in Table 4. Several alloys were tested for their effect onlaser strip welds in a 680° F. water autoclave test for an 84 dayperiod. Some of these alloys had chromium, while the other alloys didnot include chromium except in unintentional trace amounts. Still otheralloy tube welds were tested in the form of magnetic force welds in an879-day 680° F. water autoclave test. Each weld specimen placed in thetwo autoclave tests contained the weld and about 0.25 inches of an endplug and tube on either side of the weld. Separate same length tubespecimens without the weld were also included in the test. The weightgain data were collected on the weld and tube specimens. The ratio ofthe weld corrosion to the non-weld corrosion was determined either fromthe weight gain data or the metallographic oxide thickness measurementsat different locations on the specimen.

TABLE 4 Weld/Base Corrosion Alloy Name Composition by weight % RatioLASER STRIP WELDS Standard ZIRLO Zr—0.95Nb—1.08Sn—0.11Fe 2.07 Zr—NbZr—1.03Nb 2.307 Low-Sn ZIRLO Zr—1.06Nb—0.73Sn—0.27Fe 1.71 StandardZr—0.97Nb—0.99Sn—0.10Fe 2.094 ZIRLO/590° C. RXA Alloy AZr—0.31Nb—0.51Sn—0.35Fe— 1.333 0.23Cr MAGNETIC FORCE TUBE WELDS OptinZr-4 Zr—1.35Sn—0.22Fe—0.10Cr 0.805 Zr-4 + Fe Zr—1.28Sn—0.33Fe—0.09Cr0.944 Zr—2P Zr—1.29Sn—0.18Fe—0.07Ni— 1.008 0.10Cr Alloy CZr—0.4Sn—0.5Fe—0.24Cr 0.955 Alloy E Zr—0.4Nb—0.7Sn—0.45Fe— 1.1680.03Ni—0.24Cr

As shown in Table 4, the ratios of the zirconium alloys not havingchromium had a weld to base metal corrosion ratio of 1.71 or greater. Incontrast, the zirconium alloys containing chromium had a maximum ratioof 1.333 or lower. The chromium additions reduce the ratio of weldcorrosion relative to that of the base metal. Thus, the addition ofchromium significantly reduces weld corrosion, thereby increasing thesafety, cost and efficiency of the nuclear fuel assembly.

The differences in weld versus base metal corrosion may be explained bydifferences in vacancy concentration. The weld region is heated to hightemperature during welding, and cools at a faster rate than the basematerial. In a typical increase of temperature, the vacancies in themetal increase exponentially with the temperature. A fraction of theatomic vacancies introduced during the temperature increase are quenchedduring the cooling of the weld and, as a result, the vacancyconcentration is higher in the weld region. Thus, the vacancyconcentration is higher in the weld than the heat affected regions ofthe non-weld region. Since waterside corrosion of zirconium alloys ispostulated to occur by vacancy exchange with oxygen ions, increasedvacancy concentration in the weld region can increase vacancy/oxygenexchange and thereby increase corrosion in the weld region if thevacancies are not pinned by an alloying element. This exchange will bereduced resulting in improvement of corrosion resistance of the weld.Due to a high solubility of chromium in beta zirconium (about 47% weightpercent), chromium is an effective solid solution element to pin thevacancies in the beta phase and thereby decrease the corrosionenhancement due to oxygen ion exchange with supersaturated vacancies inthe quenched weld region.

While a full and complete description of the invention has been setforth in accordance with the dictates of the patent statutes, it shouldbe understood that modifications can be resorted to without departingfrom the spirit hereof or the scope of the appended claims. For example,the time for the intermediate anneals can vary widely while stillmaintaining the spirit of the invention.

1. A zirconium based alloy having improved corrosion resistance, for usein an elevated temperature environment of a nuclear reactor, the alloycomprising: 0.2 to 1.5 weight percent niobium, 0.01 to 0.6 weightpercent iron, and additional alloying elements selected from the groupconsisting of: 0.0 to 0.8 weight percent tin 0.0 to 0.5 weight percentchromium 0.0 to 0.3 weight percent copper 0.0 to 0.3 weight percentvanadium 0.0 to 0.1 weight percent nickel, with the balance at least 97weight percent zirconium, including impurities, wherein said alloy ischaracterized in that it has improved corrosion resistance propertiesdue to a final heat treatment selected from one of i) SRA or PRXA(15-20% RXA) providing low corrosion resistance; or ii) RXA or PRXA(80-95% RXA) providing low creep rate.
 2. The zirconium alloy of claim1, said alloy characterized in that it has improved in-reactor corrosionresistance properties due to a final heat treatment of SRA or PRXA(15-20% RXA).
 3. The zirconium alloy of claim 1, said alloycharacterized in that it has improved in-reactor creep resistanceproperties due to a final heat treatment of RXA or PRXA (80-95% RXA). 4.The zirconium alloy of claim 1, wherein the alloy has a compositionconsisting essentially of: 0.4 to 1.5 weight percent niobium, 0.05 to0.4 weight percent tin, 0.01 to 0.1 weight percent iron, 0.02 to 0.3weight percent copper, 0.12 to 0.3 weight percent vanadium, 0.0 to 0.5weight percent chromium, the balance at least 97 weight percentzirconium, including impurities, wherein said alloy is characterized inthat it has improved corrosion resistance properties due to a final heattreatment selected from one of i) SRA or PRXA (15-20% RXA) providing lowcorrosion resistance; or ii) RXA or PRXA (80-95% RXA) providing lowcreep rate.
 5. The zirconium alloy of claim 4, wherein the alloy has acomposition of about 1.0 weight percent niobium, 0.3 weight percent tin,0.05 weight percent iron, 0.12 weight percent copper, 0.18 weightpercent vanadium, the balance at least 97 weight percent zirconium,including impurities.
 6. The zirconium alloy of claim 4, wherein thealloy has a composition of about 0.7 weight percent niobium, 0.3 weightpercent tin, 0.05 weight percent iron, 0.12 weight percent copper, 0.18weight percent vanadium, the balance at least 97 weight percentzirconium, including impurities.
 7. The zirconium alloy of claim 4, saidalloy characterized in that it has improved in-reactor corrosionresistance properties due to a final heat treatment of SRA or PRXA(15-20% RXA).
 8. The zirconium alloy of claim 4, said alloycharacterized in that it has improved in-reactor creep resistanceproperties due to final heat treatment of RXA or PRXA (80-95% RXA). 9.The zirconium alloy of claim 1, wherein the alloy has a compositionconsisting essentially of: 0.6 to 1.5 weight percent niobium, 0.01 to0.1 weight percent iron, 0.02 to 0.3 weight percent copper, 0.15 to 0.35weight percent chromium, the balance at least 97 weight percentzirconium, including impurities, wherein said alloy is characterized inthat it has improved corrosion resistance properties due to a final heattreatment selected from one of i) SRA or PRXA (15-20% RXA) providing lowcorrosion resistance; or ii) RXA or PRXA (80-95% RXA) providing lowcreep rate.
 10. The zirconium alloy of claim 9, wherein the alloy has acomposition of: 1.0 weight percent niobium, 0.05 weight percent iron,0.08 weight percent copper, 0.25 weight percent chromium, the balance atleast 97 weight percent zirconium, including impurities.
 11. Thezirconium alloy of claim 9, said alloy characterized in that it hasimproved in-reactor corrosion resistance properties due to a final heattreatment of SRA or PRXA (15-20% RXA).
 12. The zirconium alloy of claim9, said alloy characterized in that it has improved in-reactor creepresistance properties due to final heat treatment of RXA or PRXA (80-95%RXA).
 13. The zirconium alloy of claim 1, wherein the alloy has acomposition consisting essentially of: 0.2 to 1.5 weight percentniobium, 0.05 to 0.4 weight percent tin, 0.25 to 0.45 weight percentiron, 0.15 to 0.35 weight percent chromium, 0.01 to 0.1 weight percentnickel, the balance at least 97 weight percent zirconium, includingimpurities.
 14. The zirconium alloy of claim 13, wherein the alloy has acomposition of 0.7 weight percent niobium, 0.3 weight percent tin, 0.35weight percent iron, 0.25 weight percent chromium, 0.05 weight percentnickel, the balance at least 97 weight percent zirconium, includingimpurities.
 15. The zirconium alloy of claim 13, said alloycharacterized in that it has improved in-reactor corrosion resistanceproperties due to a final heat treatment of SRA or PRXA (15-20% RXA).16. The zirconium alloy of claim 13, said alloy characterized in that ithas improved in-reactor creep resistance properties due to final heattreatment of RXA or PRXA (80-95% RXA).
 17. The zirconium based alloy foruse in an elevated temperature environment of a nuclear reactor, whereinthe alloy has a composition comprising: 0.4 to 1.5 weight percentniobium, 0.4 to 0.8 weight percent tin, 0.05 to 0.3 weight percent iron,0.0 to 0.5 weight percent chromium, the balance at least 97 weightpercent zirconium, including impurities.
 18. The zirconium based alloyof claim 17, wherein the alloy has a composition of about: 0.4 to 1.5weight percent niobium, 0.6 to 0.7 weight percent tin, 0.05 to 0.3weight percent iron, the balance at least 97 weight percent zirconium,including impurities.
 19. The zirconium based alloy of claim 17, whereinthe alloy has a composition of about: 0.4 to 1.5 weight percent niobium,0.61 to 0.69 weight percent tin, 0.05 to 0.3 weight percent iron, thebalance at least 97 weight percent zirconium, including impurities. 20.The zirconium alloy of claim 17, wherein the alloy has a composition ofabout: 1.0 weight percent niobium, 0.65 weight percent tin, 0.1 weightpercent iron, the balance at least 97 weight percent zirconium,including impurities
 21. The zirconium alloy of claim 17, wherein thealloy further comprises: 0.05 to 0.5 weight percent chromium.
 22. Thezirconium alloy of claim 17, wherein the alloy has a composition ofabout: 1.0 weight percent niobium, 0.65 weight percent tin, 0.1 weightpercent iron, 0.2 weight percent chromium the balance at least 97 weightpercent zirconium, including impurities.